Comparison of two three-dimensional heterogeneous variational nodal methods for PWR control rod cusping effect and pin-by-pin calculation

2017 ◽  
Vol 101 ◽  
pp. 370-380 ◽  
Author(s):  
Yongping Wang ◽  
Hongchun Wu ◽  
Yunzhao Li
Keyword(s):  
Three Dimensional ◽  
Control Rod ◽  
Nodal Methods

Numerical Analysis of the Corium Behavior Within the Fuel Support Piece by MPS

2016 ◽  
Author(s):  
Ronghua Chen ◽  
Lie Chen ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Eutectic Reaction ◽  
Flow Behavior ◽  
Three Dimensional ◽  
Severe Accident ◽  
Guide Tube ◽  
Melt Flow ◽  
Control Rod ◽  
Mps Method ◽  
Moving Particle ◽  
Accident Conditions

In the typical boiling water reactor (BWR), each control rod guide tube supports four fuel assemblies via an orificed fuel support piece in which a channel is designed to be a potential corium relocation path from the core region to the lower head under severe accident conditions. In this study, the improved Moving Particle Semi-implicit (MPS) method was adopted to analyze the melt flow and ablation behavior in this region during a severe accident of BWR. A three-dimensional particle configuration was constructed for analyzing the melt flow behavior within the fuel support piece. Considering the symmetry of the fuel support piece, only one fourth of the fuel support was simulated. The eutectic reaction between Zr (the material of the corium) and stainless steel (the material of the fuel support piece) was taken into consideration. The typical melt flow and freezing behaviors within the fuel support piece were successfully reproduced by MPS method. In all the simulation cases, the melt discharged from the hole of the fuel support piece instead of plugging the fuel support piece. The results indicate that MPS method has the capacity to analyze the melt flow and solidification behavior in the fuel support piece.

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Wang Lianjie ◽  
Lu Di ◽  
Zhao Wenbo
Keyword(s):  
Electric Power ◽  
Supercritical Water ◽  
Transient Analysis ◽  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Water Density ◽  
Fuel Temperature ◽  
Control Rod ◽  
Reactor Protection System ◽  
Supercritical Water Cooled Reactor

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.

Analysis of ABWR Critical Control Operation by TRACG Code

2009 ◽  
Author(s):  
Yoshihiko Ishii ◽  
Kazuaki Kitou ◽  
Tomohiko Ikegawa ◽  
Shin Hasegawa ◽  
Hitoshi Ochi
Keyword(s):  
Diffusion Theory ◽  
Three Dimensional ◽  
Source Model ◽  
Control Mode ◽  
Two Phase ◽  
Control Rod ◽  
Start Up ◽  
Neutron Count ◽  
Calculation Results ◽  
Operation Timing

Hitachi utilized three-dimensional transient analysis to design and verify a critical-control mode algorithm of an automatic power regulator (APR). TRACG has a three-dimensional neutron kinetics model based on diffusion theory and a six-equation two-phase flow model. To verify the APR critical-control mode algorithm, an external-neutron-source model that makes possible to simulate a sub-critical initial core, and an APR system model were developed and added on TRACG. The code was verified by comparison of measurements and calculation results of ABWR start-up operation under the critical-control mode. The modified TRACG could simulate neutron count rates of start-up-range neutron monitors (SRNMs), reactor period, control rod operation timing, CR withdrawal length, and time of criticality declaration, well.

Experimental Analyses by SIMMER-III on Duct-Wall Failure and Fuel Discharge/Relocation Behavior

2013 ◽  
Author(s):  
Hidemasa Yamano ◽  
Yoshiharu Tobita
Keyword(s):  
Three Dimensional ◽  
Computer Code ◽  
Failure Behavior ◽  
Guide Tube ◽  
Control Rod ◽  
Duct Wall ◽  
Experimental Analyses ◽  
Multi Phase ◽  
Pool Formation ◽  
Good Agreement

This paper describes experimental analyses using SIMMER-III/IV, which are two/three-dimensional multi-component multi-phase Eulerian fluid-dynamics codes, for the purpose of the code validation. Two topics of key phenomena in core disruptive accidents were presented in this paper: duct-wall failure and fuel discharge/relocation behavior. To analyze the duct-wall failure behavior, the SCARABEE BE+3 in-pile experiments were selected. The SIMMER-III calculation was in good agreement with the overall event progression; which was characterized by coolant boiling, clad melting, fuel failure, molten pool formation, duct-wall failure, etc.; observed in the experiment. The CAMEL C6 experiment investigated the fuel discharge and relocation behavior through a simulated control rod guide tube, which is important in evaluating the neutronic reactivity. SIMMER-IV well simulated fuel-coolant interaction, sodium voiding, fuel relocation behavior observed in the experiment. These experimental analyses indicated the validity of the SIMMER-III/IV computer code for the duct wall failure and fuel discharge/relocation behavior.

Development of Three-Dimensional Neutron Kinetics Code Based on High Order Nodal Expansion Method in Hexagonal-Z Geometry

2018 ◽  
Author(s):  
Guo Chao ◽  
Liu Yu ◽  
He Hangxing ◽  
Liu Luguo ◽  
Wang Xiaoyu ◽  
...  
Keyword(s):  
Cross Sections ◽  
Three Dimensional ◽  
Expansion Method ◽  
High Order ◽  
Benchmark Problems ◽  
Weighting Method ◽  
Control Rod ◽  
Order Polynomial ◽  
Runge Kutta ◽  
Nodal Expansion Method

To solve three-dimensional kinetics problems, a high order nodal expansion method for hexagonal-z geometry (HONEM) and a Runge-Kutta (RK) method are respectively adopted to deal with the spatial and temporal problem. In the HONEM, 1D partially-integrated flux are approximated by using four order polynomial. The two order polynomial is adopted to the approximation of partially-integrated leakages. The Runge-Kutta method is adopted as a tool for dispersing the time term of 3D kinetics equation. A flux weighting method (FWM) is used for obtaining homogenized cross sections of mix node. The three-dimensional hexagonal kinetics code has been developed based on this method and tested with two benchmark problems of VVER which are the control rod ejection without any feedback and with simple adiabatic Doppler feedback. The results calculated by this code agree well with the reference results and the code is validated.

Monte Carlo Simulation of Inclined Reactivity Control Rod and Boron Poisoning for the Canadian-SCWR Supercell

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Haykel Raouafi ◽  
Guy Marleau
Keyword(s):  
Monte Carlo ◽  
Three Dimensional ◽  
Coolant Temperature ◽  
Monte Carlo Code ◽  
Control Rod ◽  
Vertical Control ◽  
The Core ◽  
Core Height ◽  
Supercell Model ◽  
Control Rods

The Canadian-SCWR is a heavy-water moderated supercritical light-water-cooled pressure tube reactor. It is fueled with CANada deuterium uranium (CANDU)-type bundles (62 elements) containing a mixture of thorium and plutonium oxides. Because the pressure tubes are vertical, the upper region of the core is occupied by the inlet and outlet headers render it nearly impossible to insert vertical control rods in the core from the top. Insertion of solid control devices from the bottom of the core is possible, but this option was initially rejected because it was judged impractical. The option that is proposed here is to use inclined control rods that are inserted from the side of the reactor and benefit from the gravitational pull exerted on them. The objective of this paper is to evaluate the neutronic performance of the proposed inclined control rods. To achieve this goal, we first develop a three-dimensional (3D) supercell model to simulate an inclined rod located between four vertical fuel cells. Simulations are performed with the SERPENT Monte Carlo code at five axial positions in the reactor to evaluate the effect of coolant temperature and density, which varies substantially with core height, on the reactivity worth of the control rods. The effect of modifying the inclination and spatial position of the control rod inside the supercell is then analyzed. Finally, we evaluate how boron poisoning of the moderator affects their effectiveness.

Safety Analysis of the Control Rod Drop Event Applying 3-D Neutronics/Thermal-Hydraulics Coupling Code SPARKLE

2009 ◽  
Author(s):  
Yuta Maruyama ◽  
Satoshi Imura ◽  
Junto Ogawa ◽  
Shuhei Miyake
Keyword(s):  
Power Distribution ◽  
Transient Analysis ◽  
Three Dimensional ◽  
Safety Margin ◽  
Safety Analysis ◽  
Thermal Hydraulics ◽  
Analysis Method ◽  
Dynamic Changes ◽  
Control Rod ◽  
Ejection Event

Mitsubishi Heavy Industries (MHI) has developed the SPARKLE code, which is a PWR plant system transient analysis code that includes a three-dimensional (3D) neutronics module coupled with a thermal-hydraulics module. MHI has performed a study of the applicability of the SPARKLE code to the events which are associated with dynamic changes in power distribution, such as the rod ejection event or the steam line break event. In this paper, MHI has applied the SPARKLE code to the control rod drop event (drop of multiple rods), which features such a power distribution change. In addition, the neutron flux detection is dependent on the location of the dropped rods in this event, which can be dynamically calculated in the SPARKLE code. By applying the SPARKLE code to the control rod drop event, it was confirmed that the safety margin for this event is sufficiently larger than the margin calculated using the current safety analysis method, even if the appropriate conservative assumptions are made.

Deceleration Performance and Parameter Influence Analysis of the Control Rod Hydraulic Deceleration Device for Control Rod Hydraulic Drive System

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Qin Benke ◽  
Li Leishi ◽  
Bo Hanliang
Keyword(s):  
Internal Control ◽  
Three Dimensional ◽  
Hydraulic Drive ◽  
Drive System ◽  
Field Analysis ◽  
Control Rod ◽  
Design And Optimization ◽  
Main Components ◽  
Technology Control ◽  
The Impact

Control rod hydraulic drive system (CRHDS), which is invented by INET, Tsinghua University, is a new type of internal control rod drive technology. Control rod hydraulic deceleration device (CRHDD), which consists of the plug, the hydraulic deceleration cylinder, etc., is one of the main components of the CRHDS. The CRHDD performs the rod dropping deceleration function through the interworking of the plug and the deceleration cylinder, which is filled with water, and reduces the rod dropping peak acceleration and the impact force acting upon the control rod to prevent the control rod cruciform blade from being deformed or damaged. The working mechanism of the CRHDD is presented and analyzed. The theoretical model of the control rod dropping process, which is based upon force analysis of the control rod during scram process, three-dimensional flow field analysis, and flow resistance calculation of the hydraulic deceleration cylinder, the kinematics and dynamics analysis of the control rod, is built whose results are compared and validated by the CRHDS scram test results. Then the model is used to analyze the influence of the key parameters, including the fuel case gap, the plug design clearance, the working temperature, etc. on the CRHDD working performance. The research results can give guidance for the design and optimization of the CRHDD.

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